The present invention relates to a process for treating a salt waste generated from the step of a molten salt electrolytic purification in dry reprocessing of a spent metallic nuclear fuel (hereinafter referred to merely as the "metallic fuel" in some cases).
The metallic fuel is one of the novel fuels for a fast breeder reactor (FBR) which has attracted attention as a fuel capable of essentially solving the problem of nuclear fuel resources.
Examples of such a metallic fuel include a ternary alloy fuel, U-Pu-Zr, and a dry reprocessing process wherein the principle of electrolytic refining is applied in order to reprocess the spent fuel (see M. Tokiwai, Y. Tanaka, Z. Yamamoto, T. Kuroki, T. Nishimura and K. Yokoo: Nuclear Engineering, Vol. 33, p. 66 (1987); M. Tokiwai: Nuclear Engineering, Vol. 34, p. 46 (1988); and L. Burris, R. K. Steunenberg and W. E. Miller: Annual AICHE Meeting, Miami, Fla., Nov. 2-7 (1986)).
Since the metallic fuel is dry reprocessed as described above, wastes peculiar to the dry reprocessing, such as molten metals and salts, are generated unlike in the wet reprocessing and various proposals have been made on the processing and disposal of the wasters.
Regarding a salt waste among these wastes, a cementation process has been proposed by Argonne National Laboratory (ANL) in U.S.A. This process comprises bringing the salt waste into contact with a liquid metal (Cd-Li) to extract transuranic elements (TRU) into cadmium, conducting group partition, mixing the partitioned material with a cement matrix material and water, injecting the mixture in a liquid state into a metallic container, and solidifying the mixture.
The cementation process, however, has problems that the leaching resistance is low and hydrogen and chlorine are generated under radiation exposure, so that the development of a method of stably and safely storing long-lived nuclides for a long period of time is desired in the art.